A wide variety of radioactive isotopes are known for use in medical, industrial, research, and commercial applications. In general, such radioisotopes are produced by nuclear fission or by irradiating target isotope materials with nuclear particles, such that the target atoms either transmute directly into the desired isotope, or a desired radioisotope is produced through a chain of absorption and decay that subsequently generates the desired radioactive product.
One particularly important radioisotope, widely used in medical imaging, is Technetium-99m (Tc-99m), which is a metastable isotope that results from beta decay of Molybdenum-99 (Mo-99). Tc-99 is a pure gamma emitter with a 6 hour half-life, emitting mono-energetic 140.5 keV photons. Mo-99 may be made either by bombarding Mo-98 with neutrons or protons, or by the only method of Mo-99 production approved by the FDA for medical use; the extraction of Mo-99 from the fission products of Uranium-235 (U-235). Tc-99m is used in approximately 80 percent of all nuclear medicine diagnostic procedures, and in roughly 50,000 diagnostic and therapeutic nuclear medicine procedures performed daily in the United States, including diagnosis of heart disease, treating cancer, and studying organ structure and function. However, the short half-life of Tc-99m requires that it be produced continuously to meet the medical community's needs.
Presently, most of the world's supply of Mo-99 for medical applications comes from only a few reactors, all of which are nearing end of life, and none of which are located in the United States. For example, Canada's National Research Universal (NRU) reactor at the Atomic Energy of Canada's (AECL's) Chalk River Laboratories in Ontario, Canada, produced a high percentage of the world's medical and industrial radioisotopes, including Mo-99. However, in November 2007, NRU was shutdown due to a heavy water leak and did not resume Mo-99 production operations until September 2010. In February 2010, the 49-year-old High Flux Reactor in the Netherlands was shutdown for a four-month maintenance period. Even routine maintenance of such facilities can cause significant disruption in the supply of Mo-99 and Tc-99m available for medical applications. Also the processing and distribution of isotopes is complex and constrained, which can be critical when the isotopes concerned are short-lived. A need for increased production capacity and more reliable distribution is evident.
Reacting to projected shortfalls in Tc-99m availability, and to address concerns about proliferation arising from the use of high-enriched uranium (HEU), the United States Congress has called for all Mo-99 to be supplied by reactors running on low-enriched uranium (LEU), e.g., having U-235 enrichments of 20% or less. Congress also has sought proposals for an LEU-based supply of Mo-99 for the US market that would supply 111 six-day Terabecquerel (TBq) per week by mid-2013.
Two different technologies currently are available to provide the source of radiation for Mo-99 production: fission nuclear reactors, which produce a flux of neutrons; and particle accelerators or cyclotrons, which produce a flux of charged particles. In one method of production, a target of natural molybdenum or molybdenum enriched in Mo-98 is irradiated by a neutron flux in a nuclear reactor. Mo-99 results from neutron radiation capture. The irradiated target then undergoes reprocessing to extract and isolate the Mo-99. This method, however, has a low yield and is characterized by a low specific activity due to presence of Mo-98 in the final product.
The only method currently approved by the U.S. Food and Drug Administration (FDA) for the production of Mo-99 consists of irradiating an enriched uranium target in a nuclear reactor and then chemically extracting Mo-99. In this method, a uranium target comprising a U—Al alloy or electroplated target, containing enriched U-235, is irradiated in a nuclear reactor to cause the U-235 to fission. Currently 93% enriched uranium is used but efforts are underway to employ 20% enriched uranium. After irradiation, the target is dissolved and the resultant solution processed by any of a number of previously-known radiochemical processing methods to extract Mo-99 from the fission products. The specific activity achieved by this method is several tens of kilocuries per gram of molybdenum. A serious drawback of this method, however, is that it generates large amounts of radioactive wastes during recovery of the Mo-99, which typically exceed the amount of recovered Mo-99 by two orders of magnitude. Also, the waste contains almost all of the U-235 originally contained in the target. In the case of highly enriched uranium, the waste presents a proliferation problem as the waste could be diverted to weapons production. In addition, such previously-known methods usually involve at least a 24-hour cool-down period prior to processing the irradiated uranium targets, during which time the Mo-99 activity decreases by at least 22%. After a two-day delay, the activity of the waste byproducts exceeds that of the Mo-99 by a factor of six or seven. The problem of long-lived fission byproduct management is therefore a major disadvantage in the production of Mo-99 by previously-known methods.
U.S. Pat. Nos. 4,017,583 and 4,123,498 disclose the use of oxygen to strip MoO3 gas from uranium oxide at very high temperatures. The oxygen/chlorine mixture of this invention forms MoO2Cl2 which becomes a gas at a much lower temperature than MoO3. The vapor pressure of MoO3 is 10−13 atmospheres at 400° C. while MoO2Cl2 is totally vaporized at this temperature.
U.S. Pat. No. 5,596,611 discloses a small, dedicated uranyl nitrate homogeneous reactor for the production of Mo-99 in which the radioactive waste products are recirculated back into the reactor. A portion of the uranyl nitrate solution from the reactor is directly siphoned off and passed through columns of alumina to fix some of the fission products, including Mo-99, to the alumina. The Mo-99 and some fission products on the alumina column then are removed through elution with a hydroxide and the Mo-99 is either precipitated out of the resultant elutriant with alpha-benzoinoxime or passed through other columns. This uranyl nitrate reactor has the advantage of recycling the fission byproducts, but the conventional extraction method to obtain Mo-99 is relatively inefficient.
U.S. Pat. No. 5,910,971 discloses the ability to produce Mo-99 directly from the uranyl sulfate solution of an aqueous-homogenous solution nuclear reactor. The solution of the reactor is pumped through a solid sorbent material that selectively absorbs the Mo-99. The uranyl sulfate and all fission products not adhering to the sorbent are returned to the reactor vessel. This invention only takes place after shutdown and following a cool-down period, during which a large percentage of the available Mo-99 decays.
In view of the above-noted drawbacks of previously-known systems, it would be desirable to provide methods and apparatus for producing and extracting Mo-99 that would enable the establishment of a robust and reliable domestic U.S. supply of Mo-99. If it can be easily accomplished, it is also desirable to extract other useful fission product isotopes at the same time such as Xe-135, Kr-85, I-133, I-135, Cs-137, Nd-147, Rh-105, Pr-143 and Pm-147 and others.
Due to the long lead times needed to obtain regulatory approval for new reactor designs, it would be particularly desirable to provide methods and apparatus for producing and extracting Mo-99 suitable for use with existing reactor facilities and modes of operation.
Further, due to the short half-lives of Mo-99 and other radioisotopes, it would be desirable to provide methods and apparatus that reduce the time required to extract the radioisotopes, and thus increase the available amounts of such material.
It still further would be desirable to provide methods and apparatus for producing and extracting Mo-99 that reduce the wastes resulting from production and processing of Mo-99, thereby reducing the burden on existing long-term waster storage facilities.
Furthermore, it would be desirable to return the target, with its U-235 content directly back to the reactor for more Mo-99 production without additional chemical processing.
It also would be desirable to provide methods and apparatus for producing and extracting Mo-99 that enable the use of low enriched uranium targets, thereby furthering the goal of reducing worldwide distribution of highly enriched uranium and lessening the prospects for diversion and proliferation.